ML20343A243

Proposed License Amendment Request Revise Reactor Core Safety Limit to Reflect WCAP-17642-P-A, Revision 1

ML20343A243
Person / Time
Site:	Millstone
Issue date:	12/08/2020
From:	Mark D. Sartain Dominion Energy Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
20-401
Download: ML20343A243 (15)

Text

Dominion Energy Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion Energy.com December 8, 2020 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 PROPOSED LICENSE AMENDMENT REQUEST Dominion Energy Serial No.20-401 NSS&L/TFO:

RO Docket No.

50-423 License No.

NPF-49 REVISE REACTOR CORE SAFETY LIMIT TO REFLECT WCAP-17642-P-A, REVISION 1 Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) is submitting a license amendment request (LAR) to revise the Technical Specifications (TS) for Millstone Power Station Unit 3 (MPS3). The proposed change revises TS 2.1.1, "Safety Limit, Reactor Core," Safety Limit 2.1.1.2 to reflect the peak fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PADS)." provides DENC's description and assessment of the proposed change. provides the marked-up MPS3 TS page to reflect the proposed amendment.

There is no associated TS Bases change.

The proposed amendment does not involve a Significant Hazards Consideration under the standards set forth in 10 CFR 50.92. The basis for this determination is included in. DENC has also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released offsite, or any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change.

The proposed amendment has 1been reviewed and approved by the station's Facility Safety Review Committee.

DENC requests approval of this license amendment request by December 31, 2021, with a 90-day implementation period.

In accordance with 10 CFR 50.91(b), a copy of this license amendment request is being provided to the State of Connecticut.

Serial No.20-401 Docket No. 50-423 Page 2 of 3 If you have any questions or require additional information, please contact Mr. Shayan Sinha at (804) 273-4687.

Sincerely,

~

Mark D. Sartain Vice President-Nuclear Engineering and Fleet Support COMMONWEAL TH OF VIRGINIA

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)

COUNTY OF HENRICO

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The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mark D. Sartain, who is Vice President - Nuclear Engineering and Fleet Support of Dominion Energy Nuclear Connecticut, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this g-J-hday of J)u,e.,wb.et", 2020.

My Commission Expires:

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AIG D SL Notary Public Commonwealth of Vi Reg.# 751865 My Commission Expire Attachments:

1. Description and Assessment of Proposed Change

2. Marked-up Technical Specification Page Commitments made in this,.letter: None

cc:

U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 Richard V. Guzman Senior Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 C2 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Millstone Power Station Director, Radiation Division Department of Energy and Environmental Protection 79 Elm Street Hartford, CT 06106-5127 Serial No.20-401 Docket No. 50-423 Page 3 of 3 Serial No.20-401 Docket No. 50-423 DESCRIPTION AND ASSESSMENT OF PROPOSED CHANGE Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3

1.0

SUMMARY

DESCRIPTION Serial No.20-401 Docket No. 50-423, Page 1 of 9 Pursuant to 10 CFR 50.90, Dominion Energy Nuclear Connecticut, Inc. (DENC) requests an amendment to the Millstone Power Station Unit 3 (MPS3) Facility Operating License number NPF-49 in the form of a change to the Technical Specifications (TS). The proposed change revises TS 2.1.1, "Safety Limits, Reactor Core," Safety Limit (SL) 2.1.1.2 to reflect the peak fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PADS)."

This attachment provides DEN C's description and assessment of the proposed change. provides the marked-up MPS3 TS page to reflect the proposed amendment.

There is no associated TS Bases change.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation MPS3 must ensure acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences, consistent with the MPS3 current licensing basis.

To accomplish this, MPS3 TS 2.1.1, "Safety Limits, Reactor Core," ensures Departure from Nucleate Boiling (DNB) does not occur and the fuel centerline temperature remains below the fuel melting temperature. The proposed amendment revises the fuel centerline melting temperature specified in SL 2.1.1.2 but does not alter the SL associated with the DNB ratio.

Fuel centerline melting occurs when the local linear heat rate (LHR), or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the fuel pellet following centerline melting could cause excessive cladding stress leading to failure of the cladding and uncontrolled release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak LHR below the level at which fuel centerline melting occurs.

• Appropriate operation of the Reactor Protection System and steam generator safety valves prevents violation of the Reactor Core SLs.

2.2 Current Technical Specification Requirement SL 2.1.1.2 defines the burn up-dependent fuel temperature below which the fuel centerline temperature must be maintained. SL 2.1.1.2 applies whenever the reactor is critical. TS 2.1.1 requires that the unit be placed in

Serial No.20-401 Docket No. 50-423, Page 2 of 9 HOT STANDBY within one hour of exceeding the Reactor Core Safety Limit. The current MPS3 SL 2.1 is indicated below:

2.1.1.2 The peak fuel centerline temperature shall be maintained less than 5080°F, decreasing by 58°F per 10,000 MWD/MTU of burnup.

2.3 Reason for the Proposed Change Plant-specific safety analyses are performed to ensure compliance with the Safety Limit is maintained. Westinghouse Performance Analysis and Design Model (PAD5) Methodology Topical Report [Reference 1] defined the fuel pellet melting limit that is included in the PAD5 methodology based on available fuel pellet material properties. The Nuclear Regulatory Commission (NRC) reviewed and approved the Westinghouse methodology and concluded that the melting limits defined in Reference 1 are acceptable.

The proposed amendment will be implemented to maintain consistency between the value in SL 2.1.1.2 and the criteria used when performing confirmatory safety analyses that rely on the NRC approved methodology in Reference 1.

2.4 Description of Proposed Change The proposed change revises the peak fuel centerline temperature specified in SL 2.1.1.2 but does not alter the Action that must be taken following a violation of the limit. The following changes are proposed to the MPS3 TS.

The current version of SL 2.1.1.2 reads:

"The peak fuel centerline temperature shall be maintained less than 5080°F, decreasing by 58°F per 10,000 MWD/MTU of burnup."

The revised version of SL 2.1.1.2 would read:

"The peak fuel centerline temperature shall be maintained less than 5080°F, decreasing by 9°F per 10,000 MWDIMTU of burnup."

A mark-up of the proposed change to TS Section 2.1 is provided in Attachment

2.

3.0 TECHNICAL EVALUATION

Serial No.20-401 Docket No. 50-423, Page 3 of 9 The principal design tool used by Westinghouse for evaluating fuel rod performance is the Performance Analysis and Design (PAD) code [Reference 1]. This computer program iteratively calculates the interrelated effects of fuel and cladding deformations including fuel densification, fuel swelling, fuel relocation, fuel rod temperatures, fill and fission gas release, and rod internal pressure as a function of time and linear power. PAD evaluates the power history of a fuel rod as a series of steady-state power levels with instantaneous jumps from one power to another. The length of the fuel rod is divided into several axial segments, and each segment is assumed to operate at a constant set of conditions over its length. Fuel densification and swelling, cladding stresses and strains, temperatures, burn up and fission gas releases are calculated separately for each axial segment and the effects are integrated to obtain the overall fission gas release and resulting internal pressure for each time step. The coolant temperature rise along the fuel rod is calculated based on the flow rate and axial power distribution, and the cladding surface temperature is determined with consideration of corrosion effects and the possibility of local boiling.

Model updates incorporated into the PADS code address the fuel and cladding performance models required for high burnup fuel design. Key fuel performance updates to the PADS models include fuel thermal conductivity degradation (TCD) with burnup, enhanced high burnup athermal fission gas release (pellet rim effects) and enhanced high burnup fission gas bubble swelling. Cladding creep and growth models are also updated to reflect high burnup cladding performance. In addition to high burnup analysis capability, a key driver for the implementation of the PADS models in fuel design is to address regulatory concerns associated with fuel thermal conductivity degradation with burnup.

The PADS models are the latest evolutions of the Westinghouse PAD code [Reference 1]. As part of the Reference 1 development, the burnup-dependent term of the fuel melting limits in PADS was updated based on journal-published fuel material data. Additional validation performed in Section 2.1 of Appendix A of Reference 1 shows that the PADS code, in conjunction with the new fuel melt limit, accurately predicts fuel melt based on comparisons to experimental observations. Section 3.7.12 of the NRC Safety Evaluation Report in Reference 1 concludes that the fuel melting limits in PADS are acceptable.

The peak fuel centerline temperature SL is independent of the PADS methodology, as noted in the Safety Evaluation of the Amendment that reflected the SL revision for Turkey Point Nuclear Plant, Units 3 and 4 [Reference 2]. The current licensing basis safety*

analyses use the existing SL 2.1.1.2 for fuel melt as an acceptance criterion as required by the current methodology. Thus, DENC will continue to meet the existing SL when using its current licensing basis safety analyses even with the implementation of the proposed SL. Since the existing SL for peak fuel centerline temperature is more restrictive than the proposed limit, the current licensing basis safety analyses remain conservative with respect to the proposed SL.

A comprehensive description of the PADS models, NRC Requests for Additional Information, and the subsequent NRC Safety Evaluation are documented in Reference

Serial No.20-401 Docket No. 50-423, Page 4 of 9

1. The NRC Safety Evaluation Limitations and Conditions are discussed in Section 3.1 of this amendment request. As described in Section 3.1, the proposed SL will only be applicable for analyses performed with the method described in Reference 1.

3.1 Limits of Applicability The proposed amendment will only be used in applicable safety analyses that are performed with the approved fuel performance methods in Reference 1. The Limitations and Conditions from the NRC Safety Evaluation in Reference 1 pertinent to this amendment request are detailed below along with details of how each is satisfied.

The NRG staff limits the applicability of the PAD5 code and methodology to the cladding, fuel, and reactor parameters listed in Section 4. 1 of the Safety Evaluation in Reference 1.

Response: DENC will apply PAD5 within the limits specified in Section 4.1 of Reference 1 for cladding, fuel, and reactor parameters to be used at MPS3.

Because these PAD5 inputs depend on the reload design, these parameters are validated on a cycle-specific basis.

The application of PA05 should at no time exceed the fuel melting temperature as calculated by PA05 due to the lack of properties for molten fuel in PAD5 and other properties such as thermal conductivity and fission gas release.

Response: DENC will limit the MPS3 peak fuel centerline temperature per this amendment request.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TS as part of the license. The TS ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public.

10 CFR 50.90 requires NRC approval for any modification to, addition to, or deletion from the plant Technical Specifications. Therefore, this activity requires NRC approval prior to making the proposed plant-specific changes included in this license amendment request.

Serial No.20-401 Docket No. 50-423, Page 5 of 9 10 CFR 50.36 requires that the TS include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements per 10 CFR 50.36(c)(3); (4) design features; and (5) administrative controls. This amendment application is related to the first category above since a change to the peak fuel centerline melt temperature Safety Limit is proposed.

10 CFR 50, Appendix A, Generic Design Criterion (GDC) 10, Reactor Design, requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. The requirements of GDC 10 are met by the restrictions of SL 2.1.1.2 that prevent overheating of the fuel and cladding by maintaining the steady state peak temperature below the level at which fuel centerline melting occurs.

The change to Technical Specification 2.1.1, "Safety Limits, Reactor Core,"

revises SL 2.1.1.2 to be consistent with the limit approved in Westinghouse Topical Report WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PAD5)," November 2017.

4.2 Precedents The proposed change to TS Safety Limit 2.1.1.2 changes the fuel centerline temperature to reflect that specified in WCAP-17642-P-A. The NRC has approved changes to the fuel centerline melt temperature based on WCAP-17642-P-A for Turkey Point Nuclear Plant, Units 3 and 4 [Reference 2].

An LAR requesting a similar change has been submitted for Surry Power Station, Units 1 & 2 [Reference 3] and is currently under NRC review. This MPS3 LAR follows a similar format and contains content similar to the Surry submittal.

4.3 No Significant Hazards Consideration Dominion Energy Nuclear Connecticut, Inc. (DENC) proposes a change to Millstone Power Station Unit 3 (MPS3) Technical Specification (TS) Safety Limit 2.1.1.2 to provide consistency with the peak fuel centerline temperature specified in WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PAD5)."

DENC has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

Serial No.20-401 Docket No. 50-423, Page 6 of 9 1. Does the proposed license amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No There are no design changes associated with the proposed amendment.

Design, material, and construction standards that were applicable prior to this amendment request will continue to be applicable.

The proposed amendment will not affect accident initiators or precursors or alter the design, conditions, and configuration of the facility, or the manner in which the plant is operated and maintained, with respect to such initiators or precursors.

Compliance with Safety Limit 2.1.1.2 is required to confirm that fuel cladding failure does not occur as a result of fuel centerline melting. The fuel centerline melt temperature limit is established to preclude centerline melting. The proposed change has been generically reviewed by the Nuclear Regulatory Commission (NRC) and found to be appropriately conservative with respect to the fuel material properties in Topical Report WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PAD5). The Limitations and Conditions from the NRC Safety Evaluation of the Topical Report pertinent to this amendment request have been satisfied for MPS3.

Accident analysis acceptance criteria will continue to be met with the proposed amendment. The proposed amendment will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The proposed amendment will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Final Safety Analysis Report (FSAR). Consequently, the applicable radiological dose acceptance criteria will continue to be met.

Therefore, it is concluded that the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No

Serial No.20-401 Docket No. 50-423, Page 7 of 9 There are no proposed design changes, nor are there any changes in the method by which any safety-related plant structures, systems, and components perform their specified safety functions. The proposed amendment will not affect the normal method of plant operation or change any operating parameters. No equipment performance requirements will be affected. The proposed amendment will not alter any assumptions made in the safety analyses.

The proposed amendment revises Safety Limit 2.1.1.2; however, the change does not involve a physical modification of the plant.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment. There will be no adverse effect or challenges imposed on any safety-related system as a result of this proposed amendment.

Therefore, it is concluded that the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The revised Safety Limit 2.1.1.2 has been calculated based on the NRC-approved methods, which ensure that the plant operates in compliance with applicable regulatory criteria. The proposed change has been generically reviewed by the NRC and found to be appropriately conservative with respect to the fuel material properties in Topical Report WCAP-17642-P-A, Revision 1. The Limitations and Conditions from the NRC Safety Evaluation of the Topical Report pertinent to this amendment request have been satisfied for MPS3.

There will be no effect on those plant systems necessary to perform protection functions.

No instrument setpoints or system response times are affected and none of the acceptance criteria for any accident analysis will be changed.

Consequently, the proposed amendment will have no impact on the radiological consequences of a design basis accident.

Serial No.20-401 Docket No. 50-423, Page 8 of 9 Therefore, it is concluded that the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above information, DENC concludes that the proposed change does not involve a significant hazards consideration, under standards set forth in 10 CFR 50.92(c), "Issuance of Amendment," and accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusion Based on the above discussions, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENT AL CONSIDERATIONS A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Westinghouse Topical Report WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PAD5)," November 2017.

2. Letter from USNRC to Mano Nazar (Florida Power & Light Company), Turkey Point Generating Unit Nos. 3 and 4 - Issuance of Amendment Nos. 288 and 282 Regarding Revised Reactor Core Safety Limit to Reflect Topical Report WCAP-17642-P-A, Revision 1 (EPID L-2018-LLA-0120)," August 15, 2019 (ADAMS Accession Number ML19031C891).

Serial No.20-401 Docket No. 50-423, Page 9 of 9

3. Letter from M. D. Sartain (Virginia Electric and Power Company) to the USNRC (Serial No.20-341 ), "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Proposed License Amendment Request, Revise Reactor Core Safety Limit to Reflect WCAP-17642-P-A, Revision 1," September 30, 2020 (ADAMS Accession Number ML20274A329).

MARKED-UP TECHNICAL SPECIFICATIONS PAGE Dominion Energy Nuclear Connecticut, Inc.

Millstone Power Station Unit 3 Serial No.20-401 Docket No. 50-423

-FOR IIIFORMATIOII OlllY-Serial No.20-401 Docket No. 50-423, Page 1 of 1 Au~l5t 12. 200&

2.0 SAFETY LThillTS AND illfiTING SAFETY SYSTEM SETTINGS 2.1 SAFETYLIMITS REACTOR CORE

'.U.1 The combination of1HERMAL POWER, Reactor Coolant System highest loop average t~nperature, and pressurizer pressure shall not exceed the limits specified in the CORE OPERATING LThillTS REPORT; and the following Safety Limits shall not be exceeded:

2. LU The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to Ll4 for the \\\\lRB-2MDNB correlations.

+

2. Ll.2 The peak fuel centerline temperature sh.111 be maintained less than 5080F, decreasing by 3S::f: per 10,00011,fWD/MTU ofbumup.

l9°FI APPLICABILITY:

MODES 1 and 2.

ACTION:

Whenever the Reactor Core Safety Limit is violated, restore compliance and be in HOT STA..NDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

A]>PI,IC'ABILITY-MODES 1. 2, 3. 4. and 5.

ACTION:

MODES 1 and 2:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5:

Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 nlinutes.

MIT.LSTONE - UNIT 3 2-1 Amendment No.~,~.~,

242